Validation Of A Relap5 Computer Model For A Vver 1000 Nuclear Power Plant

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Validation of a RELAP5 Computer Model for a VVER-1000 Nuclear Power Plant

Author :
Publisher :
ISBN 13 :
Total Pages : 22 pages
Book Rating : 4.:/5 (684 download)

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Book Synopsis Validation of a RELAP5 Computer Model for a VVER-1000 Nuclear Power Plant by :

Download or read book Validation of a RELAP5 Computer Model for a VVER-1000 Nuclear Power Plant written by and published by . This book was released on 1997 with total page 22 pages. Available in PDF, EPUB and Kindle. Book excerpt: This paper describes a computer model that has been developed for a VVER-1000 nuclear power plant for use with the RELAP5/MOD3.1.1 computer code in the analyses of operational occurrences abnormal events, and design basis scenarios. This model will provide a significant analytical capability for the Bulgarian nuclear regulatory body (Committee on the Use of Atomic Energy For Peaceful Purposes) and the Bulgarian technical specialists located at the power plant site (Kozloduy Nuclear Power Plant). In addition, the initial validation of computer model has been completed and is described in the paper. The analytical results are compared with data obtained during planned testing at the power plant; the test performed was the trip of a single main coolant pump. In addition, the paper provides a discussion of various other RELAP5 parameters calculated for the main coolant pump trip scenario. This model development and validation analysis represents an important accomplishment in the analyses of Russian designed nuclear power plants with computer codes developed and used in Western countries. The results indicate that RELAP5 can predict the thermal-hydraulic behavior of the VVER-1000 reactor for the class of transients represented by test results. 8 refs., 11 figs., 1 tab.

RELAP5

Author :
Publisher :
ISBN 13 :
Total Pages : 124 pages
Book Rating : 4.:/5 (684 download)

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Book Synopsis RELAP5 by :

Download or read book RELAP5 written by and published by . This book was released on 2002 with total page 124 pages. Available in PDF, EPUB and Kindle. Book excerpt: The RELAP5/MOD3.2 computer program has been used to analyze a series of tests investigating heat-transfer from a partly uncovered VVER-1000 core in the KS test facility at the Russian Research Center ''Kurchatov Institute'' (RRC-KI). The analysis documented represents VVER Standard Problem 4 in Joint Project 6, which is the investigation of Computer Code Validation for Transient Analysis of RBMK and VVER Reactors, between the United States and Russian International Nuclear Safety Centers. The experiment facility and data, RELAP5 nodalization, and results are shown for one of the six tests defined in Standard Problem 4. Only part of the data was analyzed due to our conclusion that the available experimental data is not sufficient to allow the modeling of the actual experiment sequence. The experiment initial conditions were reached through a series of transient processes, about which no quantitative information was available. This has required the modeling of an arbitrary computational transient, with the goal of reaching initial conditions similar to those observed during the experiment. The use of an arbitrary transient introduces many degrees of freedom in the analysis, i.e. initial computational values that influence the entire sequence of events, including the loop behavior during the experiment time window. Reasonable agreement between RELAP5 and the experiment data can be obtained by manipulating a number of initial computational values, including the liquid level in the fuel assembly model, the liquid level in the annular region, the quality of the saturated vapor in the voided loop regions, etc. Our study has focused on exploring the sensitivity of results to changes in these initial values which are not based on experimental information, but are selected with the goal of matching the experimentally observed behavior during the experiment time window. We have shown that changes in several initial arbitrary values can lead to similar changes in the calculated loop behavior. Other parameters exist which can directly influence the calculated results, which have been only minimally explored, or have not been explored at all due to lack of time. Because we are not modeling the entire sequence of events which led to the initial experiment conditions, it is not possible to assume that by matching a limited set of measured parameters at the beginning of the experiment time window we have reproduced all the initial actual conditions. We conclude that the analysis of these experiments, while providing a useful demonstration of the RELAP5 capabilities to describe the experimental sequence of events, cannot be used to draw quantitative conclusions about the adequacy of specific RELAP5 models.

Extremely Accurate Sequential Verification of RELAP5-3D.

Author :
Publisher :
ISBN 13 :
Total Pages : 25 pages
Book Rating : 4.:/5 (965 download)

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Book Synopsis Extremely Accurate Sequential Verification of RELAP5-3D. by :

Download or read book Extremely Accurate Sequential Verification of RELAP5-3D. written by and published by . This book was released on 2015 with total page 25 pages. Available in PDF, EPUB and Kindle. Book excerpt: Large computer programs like RELAP5-3D solve complex systems of governing, closure and special process equations to model the underlying physics of nuclear power plants. Further, these programs incorporate many other features for physics, input, output, data management, user-interaction, and post-processing. For software quality assurance, the code must be verified and validated before being released to users. For RELAP5-3D, verification and validation are restricted to nuclear power plant applications. Verification means ensuring that the program is built right by checking that it meets its design specifications, comparing coding to algorithms and equations and comparing calculations against analytical solutions and method of manufactured solutions. Sequential verification performs these comparisons initially, but thereafter only compares code calculations between consecutive code versions to demonstrate that no unintended changes have been introduced. Recently, an automated, highly accurate sequential verification method has been developed for RELAP5-3D. The method also provides to test that no unintended consequences result from code development in the following code capabilities: repeating a timestep advancement, continuing a run from a restart file, multiple cases in a single code execution, and modes of coupled/uncoupled operation. In conclusion, mathematical analyses of the adequacy of the checks used in the comparisons are provided.

Improving the Safety of Soviet-designed Nuclear Power Plants

Author :
Publisher :
ISBN 13 :
Total Pages : 114 pages
Book Rating : 4.:/5 (31 download)

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Book Synopsis Improving the Safety of Soviet-designed Nuclear Power Plants by :

Download or read book Improving the Safety of Soviet-designed Nuclear Power Plants written by and published by . This book was released on 1996 with total page 114 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors

Author : Francesco D'Auria
Publisher : Elsevier
ISBN 13 : 032385611X
Total Pages : 1012 pages
Book Rating : 4.3/5 (238 download)

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Book Synopsis Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors by : Francesco D'Auria

Download or read book Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors written by Francesco D'Auria and published by Elsevier. This book was released on 2024-07-29 with total page 1012 pages. Available in PDF, EPUB and Kindle. Book excerpt: Handbook on Thermal Hydraulics of Water-Cooled Nuclear Reactors, Volume 2, Modelling includes all new chapters which delve deeper into the topic, adding context and practical examples to help readers apply learnings to their own setting. Topics covered include experimental thermal-hydraulics and instrumentation, numerics, scaling and containment in thermal-hydraulics, as well as a title dedicated to good practices in verification and validation. This book will be a valuable reference for graduate and undergraduate students of nuclear or thermal engineering, as well as researchers in nuclear thermal-hydraulics and reactor technology, engineers working in simulation and modeling of nuclear reactors, and more. In addition, nuclear operators, code developers and safety engineers will also benefit from the practical guidance provided. - Presents a comprehensive analysis on the connection between nuclear power and thermal hydraulics - Includes end-of-chapter questions, quizzes and exercises to confirm understanding and provides solutions in an appendix - Covers applicable nuclear reactor safety considerations and design technology throughout

Systems, Decision and Control in Energy I

Author : Vitaliy Babak
Publisher : Springer Nature
ISBN 13 : 3030485838
Total Pages : 281 pages
Book Rating : 4.0/5 (34 download)

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Book Synopsis Systems, Decision and Control in Energy I by : Vitaliy Babak

Download or read book Systems, Decision and Control in Energy I written by Vitaliy Babak and published by Springer Nature. This book was released on 2020-07-01 with total page 281 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book examines the problems in the field of energy and related areas (including chemistry, transport, aerospace, construction, metallurgy and engineering) that Ukrainian scientists are currently investigating. The research presented focuses on ensuring the operational reliability, durability and safety of energy equipment, as well as the development of control, diagnostics and monitoring systems in the energy sector. Further, the book explores the ecological consequences of energy facilities , particularly environmental pollution in large cities and industrial areas. Written mainly by representatives of the Council of Young Scientists of the Department of Physical and Technical Problems of Energy at the NAS of Ukraine, it is intended for researchers and engineers, as well as lecturers and postgraduates at higher education institutions interested in the control, diagnosis and monitoring of energy facilities.

Nuclear Power Plant Design and Analysis Codes

Author : Jun Wang
Publisher : Woodhead Publishing
ISBN 13 : 0128181915
Total Pages : 612 pages
Book Rating : 4.1/5 (281 download)

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Book Synopsis Nuclear Power Plant Design and Analysis Codes by : Jun Wang

Download or read book Nuclear Power Plant Design and Analysis Codes written by Jun Wang and published by Woodhead Publishing. This book was released on 2020-11-10 with total page 612 pages. Available in PDF, EPUB and Kindle. Book excerpt: Nuclear Power Plant Design and Analysis Codes: Development, Validation, and Application presents the latest research on the most widely used nuclear codes and the wealth of successful accomplishments which have been achieved over the past decades by experts in the field. Editors Wang, Li,Allison, and Hohorst and their team of authors provide readers with a comprehensive understanding of nuclear code development and how to apply it to their work and research to make their energy production more flexible, economical, reliable and safe.Written in an accessible and practical way, each chapter considers strengths and limitations, data availability needs, verification and validation methodologies and quality assurance guidelines to develop thorough and robust models and simulation tools both inside and outside a nuclear setting. This book benefits those working in nuclear reactor physics and thermal-hydraulics, as well as those involved in nuclear reactor licensing. It also provides early career researchers with a solid understanding of fundamental knowledge of mainstream nuclear modelling codes, as well as the more experienced engineers seeking advanced information on the best solutions to suit their needs. - Captures important research conducted over last few decades by experts and allows new researchers and professionals to learn from the work of their predecessors - Presents the most recent updates and developments, including the capabilities, limitations, and future development needs of all codes - Incudes applications for each code to ensure readers have complete knowledge to apply to their own setting

RELAP[five]/MOD[three Point Two] Analysis of a VVER-[one Thousand] Reactor with UO[subscript Two] Fuel and MOX Fuel

Author : Chun Fu
Publisher :
ISBN 13 :
Total Pages : 128 pages
Book Rating : 4.:/5 (466 download)

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Book Synopsis RELAP[five]/MOD[three Point Two] Analysis of a VVER-[one Thousand] Reactor with UO[subscript Two] Fuel and MOX Fuel by : Chun Fu

Download or read book RELAP[five]/MOD[three Point Two] Analysis of a VVER-[one Thousand] Reactor with UO[subscript Two] Fuel and MOX Fuel written by Chun Fu and published by . This book was released on 2000 with total page 128 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Transactions of the American Nuclear Society

Author : American Nuclear Society
Publisher :
ISBN 13 :
Total Pages : 1096 pages
Book Rating : 4.3/5 (91 download)

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Book Synopsis Transactions of the American Nuclear Society by : American Nuclear Society

Download or read book Transactions of the American Nuclear Society written by American Nuclear Society and published by . This book was released on 2002 with total page 1096 pages. Available in PDF, EPUB and Kindle. Book excerpt:

RELAP-7 Software Verification and Validation Plan

Author :
Publisher :
ISBN 13 :
Total Pages : 176 pages
Book Rating : 4.:/5 (962 download)

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Book Synopsis RELAP-7 Software Verification and Validation Plan by :

Download or read book RELAP-7 Software Verification and Validation Plan written by and published by . This book was released on 2014 with total page 176 pages. Available in PDF, EPUB and Kindle. Book excerpt: This INL plan comprehensively describes the software for RELAP-7 and documents the software, interface, and software design requirements for the application. The plan also describes the testing-based software verification and validation (SV & V) process--a set of specially designed software models used to test RELAP-7. The RELAP-7 (Reactor Excursion and Leak Analysis Program) code is a nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on the INL's modern scientific software development framework - MOOSE (Multi-Physics Object-Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5's capability and extends the analysis capability for all reactor system simulation scenarios.

Nuclear Power

Author : Pavel Tsvetkov
Publisher : BoD – Books on Demand
ISBN 13 : 9533075066
Total Pages : 208 pages
Book Rating : 4.5/5 (33 download)

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Book Synopsis Nuclear Power by : Pavel Tsvetkov

Download or read book Nuclear Power written by Pavel Tsvetkov and published by BoD – Books on Demand. This book was released on 2011-09-06 with total page 208 pages. Available in PDF, EPUB and Kindle. Book excerpt: At the onset of the 21st century, we are searching for reliable and sustainable energy sources that have a potential to support growing economies developing at accelerated growth rates, technology advances improving quality of life and becoming available to larger and larger populations. The quest for robust sustainable energy supplies meeting the above constraints leads us to the nuclear power technology. Today's nuclear reactors are safe and highly efficient energy systems that offer electricity and a multitude of co-generation energy products ranging from potable water to heat for industrial applications. Catastrophic earthquake and tsunami events in Japan resulted in the nuclear accident that forced us to rethink our approach to nuclear safety, requirements and facilitated growing interests in designs, which can withstand natural disasters and avoid catastrophic consequences. This book is one in a series of books on nuclear power published by InTech. It consists of ten chapters on system simulations and operational aspects. Our book does not aim at a complete coverage or a broad range. Instead, the included chapters shine light at existing challenges, solutions and approaches. Authors hope to share ideas and findings so that new ideas and directions can potentially be developed focusing on operational characteristics of nuclear power plants. The consistent thread throughout all chapters is the "system-thinking" approach synthesizing provided information and ideas. The book targets everyone with interests in system simulations and nuclear power operational aspects as its potential readership groups - students, researchers and practitioners.

Proceedings of the 2nd ASME-JSME Nuclear Engineering Joint Conference

Author : P. F. Peterson
Publisher :
ISBN 13 :
Total Pages : 776 pages
Book Rating : 4.F/5 ( download)

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Book Synopsis Proceedings of the 2nd ASME-JSME Nuclear Engineering Joint Conference by : P. F. Peterson

Download or read book Proceedings of the 2nd ASME-JSME Nuclear Engineering Joint Conference written by P. F. Peterson and published by . This book was released on 1993 with total page 776 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Risk, Reliability and Safety: Innovating Theory and Practice

Author : Lesley Walls
Publisher : CRC Press
ISBN 13 : 1315349167
Total Pages : 4767 pages
Book Rating : 4.3/5 (153 download)

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Book Synopsis Risk, Reliability and Safety: Innovating Theory and Practice by : Lesley Walls

Download or read book Risk, Reliability and Safety: Innovating Theory and Practice written by Lesley Walls and published by CRC Press. This book was released on 2016-11-25 with total page 4767 pages. Available in PDF, EPUB and Kindle. Book excerpt: The safe and reliable performance of many systems with which we interact daily has been achieved through the analysis and management of risk. From complex infrastructures to consumer durables, from engineering systems and technologies used in transportation, health, energy, chemical, oil, gas, aerospace, maritime, defence and other sectors, the management of risk during design, manufacture, operation and decommissioning is vital. Methods and models to support risk-informed decision-making are well established but are continually challenged by technology innovations, increasing interdependencies, and changes in societal expectations. Risk, Reliability and Safety contains papers describing innovations in theory and practice contributed to the scientific programme of the European Safety and Reliability conference (ESREL 2016), held at the University of Strathclyde in Glasgow, Scotland (25—29 September 2016). Authors include scientists, academics, practitioners, regulators and other key individuals with expertise and experience relevant to specific areas. Papers include domain specific applications as well as general modelling methods. Papers cover evaluation of contemporary solutions, exploration of future challenges, and exposition of concepts, methods and processes. Topics include human factors, occupational health and safety, dynamic and systems reliability modelling, maintenance optimisation, uncertainty analysis, resilience assessment, risk and crisis management.

Thermal-Hydraulics of Water Cooled Nuclear Reactors

Author : Francesco D'Auria
Publisher : Woodhead Publishing
ISBN 13 : 0081006799
Total Pages : 1200 pages
Book Rating : 4.0/5 (81 download)

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Book Synopsis Thermal-Hydraulics of Water Cooled Nuclear Reactors by : Francesco D'Auria

Download or read book Thermal-Hydraulics of Water Cooled Nuclear Reactors written by Francesco D'Auria and published by Woodhead Publishing. This book was released on 2017-05-18 with total page 1200 pages. Available in PDF, EPUB and Kindle. Book excerpt: Thermal Hydraulics of Water-Cooled Nuclear Reactors reviews flow and heat transfer phenomena in nuclear systems and examines the critical contribution of this analysis to nuclear technology development. With a strong focus on system thermal hydraulics (SYS TH), the book provides a detailed, yet approachable, presentation of current approaches to reactor thermal hydraulic analysis, also considering the importance of this discipline for the design and operation of safe and efficient water-cooled and moderated reactors. Part One presents the background to nuclear thermal hydraulics, starting with a historical perspective, defining key terms, and considering thermal hydraulics requirements in nuclear technology. Part Two addresses the principles of thermodynamics and relevant target phenomena in nuclear systems. Next, the book focuses on nuclear thermal hydraulics modeling, covering the key areas of heat transfer and pressure drops, then moving on to an introduction to SYS TH and computational fluid dynamics codes. The final part of the book reviews the application of thermal hydraulics in nuclear technology, with chapters on V&V and uncertainty in SYS TH codes, the BEPU approach, and applications to new reactor design, plant lifetime extension, and accident analysis. This book is a valuable resource for academics, graduate students, and professionals studying the thermal hydraulic analysis of nuclear power plants and using SYS TH to demonstrate their safety and acceptability. - Contains a systematic and comprehensive review of current approaches to the thermal-hydraulic analysis of water-cooled and moderated nuclear reactors - Clearly presents the relationship between system level (top-down analysis) and component level phenomenology (bottom-up analysis) - Provides a strong focus on nuclear system thermal hydraulic (SYS TH) codes - Presents detailed coverage of the applications of thermal-hydraulics to demonstrate the safety and acceptability of nuclear power plants

Energy Research Abstracts

Author :
Publisher :
ISBN 13 :
Total Pages : 1696 pages
Book Rating : 4.U/5 (183 download)

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Download or read book Energy Research Abstracts written by and published by . This book was released on 1982 with total page 1696 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Relap5-3d Model Validation and Benchmark Exercises for Advanced Gas Cooled Reactor Application

Author : Eugene James Thomas Moore
Publisher :
ISBN 13 :
Total Pages : pages
Book Rating : 4.:/5 (728 download)

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Book Synopsis Relap5-3d Model Validation and Benchmark Exercises for Advanced Gas Cooled Reactor Application by : Eugene James Thomas Moore

Download or read book Relap5-3d Model Validation and Benchmark Exercises for Advanced Gas Cooled Reactor Application written by Eugene James Thomas Moore and published by . This book was released on 2006 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: High-temperature gas-cooled reactors (HTGR) are passively safe, efficient, and economical solutions to the world's energy crisis. HTGRs are capable of generating high temperatures during normal operation, introducing design challenges related to material selection and reactor safety. Understanding heat transfer and fluid flow phenomena during normal and transient operation of HTGRs is essential to ensure the adequacy of safety features, such as the reactor cavity cooling system (RCCS). Modeling abilities of system analysis codes, used to develop an understanding of light water reactor phenomenology, need to be proven for HTGRs. RELAP5-3D v2.3.6 is used to generate two reactor plant models for a code-to-code and a code-to-experiment benchmark problem. The code-to-code benchmark problem models the Russian VGM reactor for pressurized and depressurized pressure vessel conditions. Temperature profiles corresponding to each condition are assigned to the pressure vessel heat structure. Experiment objectives are to calculate total thermal energy transferred to the RCCS for both cases. Qualitatively, RELAP5-3D's predictions agree closely with those of other system codes such as MORECA and Thermix. RELAP5-3D predicts that 80% of thermal energy transferred to the RCCS is radiant. Quantitatively, RELAP5-3D computes slightly higher radiant and convective heat transfer rates than other system analysis codes. Differences in convective heat transfer rate arise from the type and usage of convection models. Differences in radiant heat transfer stem from the calculation of radiation shape factors, also known as view or configuration factors. A MATLAB script employs a set of radiation shape factor correlations and applies them to the RELAP5-3D model. This same script is used to generate radiation shape factors for the code-to-experiment benchmark problem, which uses the Japanese HTTR reactor to determine temperature along the outside of the pressure vessel. Despite lacking information on material properties, emissivities, and initial conditions, RELAP5-3D temperature trend predictions closely match those of other system codes. Compared to experimental measurements, however, RELAP5-3D cannot capture fluid behavior above the pressure vessel. While qualitatively agreeing over the pressure vessel body, RELAP5-3D predictions diverge from experimental measurements elsewhere. This difference reflects the limitations of using a system analysis code where computational fluid dynamics codes are better suited.

Assessment of RELAP5-3D Multi-dimensional Component Model Using Data from LOFT Test L2-5

Author :
Publisher :
ISBN 13 :
Total Pages : 15 pages
Book Rating : 4.:/5 (16 download)

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Book Synopsis Assessment of RELAP5-3D Multi-dimensional Component Model Using Data from LOFT Test L2-5 by :

Download or read book Assessment of RELAP5-3D Multi-dimensional Component Model Using Data from LOFT Test L2-5 written by and published by . This book was released on 1998 with total page 15 pages. Available in PDF, EPUB and Kindle. Book excerpt: The capability of the RELAP5-3D computer code to perform multi-dimensional analysis of a pressurized water reactor (PWR) was assessed using data from the Loss-of-Fluid Test (LOFT) L2-5 experiment. The LOFT facility was a 50 MW PWR that was designed to simulate the response of a commercial PWR during a loss-of-coolant accident (LOCA). Test L2-5 simulated a 200% double-ended cold leg break with an immediate primary coolant pump trip. A three-dimensional model of the LOFT reactor vessel was developed. Calculations of the LOFT L21-5 experiment were performed using the RELAP5-3D computer code. The calculations simulated the blowdown, refill, and reflood portions of the transient. The calculated thermal-hydraulic response of the primary coolant system was generally in reasonable agreement with the test. The calculated results were also generally as good as or better than those obtained previously with RELAP5/MOD3.